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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A mechanistic study on the synthesis of β-Sialon whiskers from coal fly ash

Graphical abstract: The appearance of bead-like whiskers indicated that the growth mechanism of the β-Sialon whiskers was different from the conventional one, in which a chain of droplets were formed and then consumed to participate in the formation of the whiskers. - Highlights: • β-Sialon whiskers were synthesized using waste fly ash by carbothemal reduction reaction under nitrogen atmosphere. • Rod-like β-Sialon whiskers with a diameter of 100–500 nm were formed. • Bead-like whiskers as intermediate morphology of the growing β-Sialon whiskers were found with increasing sintering time. • The growth mechanism of β-Sialon whiskers was different from the conventional VLS mechanism. • A chain of droplets were formed and participated in the formation of the whiskers. - Abstract: β-Sialon whiskers were produced at 1420 °C through carbothemal reduction reaction under nitrogen atmosphere using fly ash from coal-fired power plants. The effects of sintering time on the phase formation and morphology of the products were investigated by X-ray diffraction (XRD), scanning electron microscope (SEM) and energy dispersive spectrometer (EDS) techniques. Rod-like β-Sialon whiskers with the diameter of 100–500 nm were successfully formed. With increasing sintering time, bead-like morphology during the growth process of the whiskers was found, and growth mechanism of β-Sialon whiskers was also discussed in detail. The growth mechanism proposed in this study was different from the conventional vapor–liquid–solid (VLS) mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2021

CO2 Capture Strategies via Mineralization with Industrial Waste Brines

Large coal-fired power plants (>500 MW) account for 30% of global CO2 emissions, and long-term management of this CO2 to is urgently needed mitigate global temperature increases. Sequestration of CO2 within stable mineral carbonates (e.g., CaCO3) represents an attractive emission reduction strategy because it offers a leakage-free alternative to geological storage of CO2 in an environmentally friendly form. We have previously described a mineralization process in which divalent cations are sourced from various waste streams (e.g., produced water and brackish water) and alkalinity is induced via regenerable ion-exchange materials (Bustillos et. al. Frontiers in Energy Research. 2020, 8, 352). In our process, aqueous carbonate-bearing streams with pH > 8 are produced by contacting fresh water and carbon dioxide with various ion-exchange materials (e.g., Na form zeolites or ion exchange resins). These streams are mixed with produced water containing varying concentrations (~0.01 – 1.0 M) of Ca2+ leading to the precipitation of solid calcium carbonate (PCC). This process has the advantages of using regenerable solids in a simple and continuous process to increase the pH of water by ion exchange instead of relying on the consumption of costly and unsustainable sources of alkalinity (e.g., sodium hydroxide). While once-through column experiments showed the above benefits, the same were yet to established in a steady-state process with recycle streams. In this work, we set up a process simulation to quantify the energy requirements and CO2 emissions associated with the process and seek optimal produced water compositions and CO2 concentrations (5 – 20 vol%). The process simulation was set up in ASPEN Plus using eRNTL as the thermodynamic property method and sequential modular strategy. Ion exchange alkaline solution was simulated using sodium hydroxide and validated against the experimental data obtained from once-through kinetic experiments. Nanofiltration and reverse osmosis membrane steps were also implemented for the separation of divalent cations and production of fresh water and a regeneration stream following mineralization. Sensitivity analysis was carried out using a range of produced water compositions (0.01 – 1.0 M Ca2+, 0.001 – 0.15 M Mg2+, 0.5 – 3.5 M Na+ and 0.0004 – 0.002 M Fe2+) according to the United States Geological Survey (USGS) database. Calcium carbonate yields increased with increasing CO2 concentrations and were maximized using produced water compositions with larger Ca2+ concentrations. Maximum calcium carbonate yields produced at 5 vol%, 12 vol% and 20 vol% CO2 were 2.3 mmol/L, 5.5 mmol/L, and 9.3 mmol/L, respectively, with the formation of brucite (a magnesium hydroxide phase, Mg(OH)2) and goethite (an iron hydroxide phase, FeOOH) as the primary contaminant phases (99% calcite, 0.6% brucite, 0.4% goethite), which agree with phases detected by XRD experimentally. These results indicate high purity calcium carbonate can be precipitated using industrial waste streams. Consequentially, energy consumption and net CO2 emissions were minimized where precipitated calcium carbonate was maximized for all produced water compositions and CO2 concentrations. Minimum energy consumptions were 0.21 kWh/ton CO2 processed, with 98% of the energy input required coming from the membrane filtration steps. Produced water compositions with large Na+ concentrations (> 0.5 M) were effective at reducing energy consumptions due to faster regeneration time of ion exchange materials. Additionally, calculated net CO2 emissions were negative for the process and ranged from -0.02 kg/ton CO2 to -0.15 kg/ton CO2 processed, indicating a low emission process. We will also present techno-economic assessment showing the economic benefits of the current process as an alternative to the addition of stoichiometric bases to induce alkalinity for the precipitation of CaCO3.

Simonetti, Dante↗
DOE OSTI2012

Chemical Differences Between Sludge Solids at the F and H Area Tank Farms

The primary source of waste solids received into the F Area Tank Farm (FTF) was from PUREX processing performed to recover uranium and plutonium from irradiated depleted uranium targets. In contrast, two primary sources of waste solids were received into the H Area Tank Farm (HTF): a) waste from PUREX processing; and b) waste from H-modified (HM) processing performed to recover uranium and neptunium from burned enriched uranium fuel. Due to the differences between the irradiated depleted uranium targets and the burned enriched uranium fuel, the average compositions of the F and H Area wastes are markedly different from one another. Both F and H Area wastes contain significant amounts of iron and aluminum compounds. However, because the iron content of PUREX waste is higher than that of HM waste, and the aluminum content of PUREX waste is lower than that of HM waste, the iron to aluminum ratios of typical FTF waste solids are appreciably higher than those of typical HTF waste solids. Other constituents present at significantly higher concentrations in the typical FTF waste solids include uranium, nickel, ruthenium, zinc, silver, cobalt and copper. In contrast, constituents present at significantly higher concentrations in the typical HTF waste solids include mercury, thorium, oxalate, and radionuclides U-233, U-234, U-235, U-236, Pu-238, Pu-242, Cm-244, and Cm-245. Because of the higher concentrations of Pu-238 in HTF, the long-term concentrations of Th-230 and Ra-226 (from Pu-238 decay) will also be higher in HTF. The uranium and plutonium distributions of the average FTF waste were found to be consistent with depleted uranium and weapons grade plutonium, respectively (U-235 comprised 0.3 wt% of the FTF uranium, and Pu-240 comprised 6 wt% of the FTF plutonium). In contrast, at HTF, U-235 comprised 5 wt% of the uranium, and Pu-240 comprised 17 wt% of the plutonium, consistent with enriched uranium and high burn-up plutonium. X-ray diffraction analyses of various FTF and HTF samples indicated that the primary crystalline compounds of iron in sludge solids are Fe 2 O 3 , Fe 3 O 4 , and FeO(OH), and the primary crystalline compounds of aluminum are Al(OH) 3 and AlO(OH). Also identified were carbonate compounds of calcium, magnesium, and sodium; a nitrated sodium aluminosilicate; and various uranium compounds. Consistent with expectations, oxalate compounds were identified in solids associated with oxalic acid cleaning operations. The most likely oxidation states and chemical forms of technetium are assessed in the context of solubility, since technetium-99 is a key risk driver from an environmental fate and transport perspective. The primary oxidation state of technetium in SRS sludge solids is expected to be Tc(IV). In salt waste, the primary oxidation state is expected to be Tc(VII). The primary form of technetium in sludge is expected to be a hydrated technetium dioxide, TcO 2 ⋅ xH 2 O, which is relatively insoluble and likely co-precipitated with iron. In salt waste solutions, the primary form of technetium is expected to be the very soluble pertechnetate anion, TcO 4 - . The relative differences between the F and H Tank Farm waste provide a basis for anticipating differences that will occur as constituents of FTF and HTF waste residue enter the environment over the long-term future. If a constituent is significantly more dominant in one of the Tank Farms, its long-term environmental contribution will likely be commensurately higher, assuming the environmental transport conditions of the two Tank Farms share some commonality. It is in this vein that the information cited in this document is provided - for use during the generation, assessment, and validation of Performance Assessment modeling results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Simulant Development of Potential 200 West Area Waste Feeds

Preliminary planning for retrieval, qualification, and pretreatment of waste in Hanford’s 200 West Area (200W) has begun as part of the West Area Risk Management project. Experimental studies to technically mature pretreatment process operations will likely be needed because of the uniqueness of 200W waste. Pacific Northwest National Laboratory formulated five simulants to represent 200W-qualified feed based on the preliminary flowsheet provided by Washington River Protection Solutions, LLC. The simulant recipes were devised using applicable historical information as a reference point to support the use of the flowsheet waste vectors, which were combined into five distinct groups. These five groups formed the basis for the liquid composition targets that were adapted into recipes using charged-balanced salt species. The liquid phase recipes were batched in 1-L quantities and analyzed at Pacific Northwest National Laboratory. Once confirmed to be stable, the liquid solutions were tested for compatibility with candidate solid components. Specific solid components were recommended based on cross-examining the proposed solid phases in the flowsheet with relevant data from the literature. Mixtures of solid components were added to aliquots of the liquid batches and sub-sampled to measure particle size distribution. The measured distribution was compared to independently created benchmark distributions appropriate for each simulant. This process was iterated until a solid phase composition that resulted in a representative particle size distribution was found. After the final compositions were confirmed, a suite of chemical and physical characterization data was collected. This report describes the simulant basis, formulation methodology, laboratory measurements, and data collected for the recipes recommended to represent 200W waste feeds.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Structural analysis of some sodium and alumina rich high-level nuclear waste glasses

Sodium and aluminum rich high level nuclear waste glasses are prone to nepheline (NaAlSiO4) crystallization. Since nepheline removes three moles of glass forming oxides (Al2O3 and SiO2) per each mole of Na2O, its formation can result in sever deterioration of the chemical durability. The present study aims at investigating the relationships between the molecular-level structure of sodium alumino-borosilicate based simulated high-level nuclear waste glasses and their crystallization behavior by infrared spectroscopy (FTIR) and x-ray diffraction, respectively. The molecular structure of most of the investigated glasses comprises of a mixture of Q2 and Q3 (Si) units while aluminum and boron are predominantly present in tetrahedral and trigonal coordination, respectively. The increasing boron content has been shown to suppress the nepheline formation in the glasses. The structural influence of various glass components on nepheline crystallization has been discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Technetium and iodine aqueous species immobilization and transformations in the presence of strong reductants and calcite-forming solutions: Remedial action implications

At the Hanford Site in southeastern Washington State, radionuclide (Tc-99/I-129) laden liquid wastes have been discharged to ground, resulting in vadose zone contamination, which provides a continuous source of these contaminants to groundwater. The presence of multiple contaminants increases the complexity of finding viable remediation technologies to sequester vadose zone contaminants in situ and protect groundwater. Although previous studies have shown the efficiency of zero valent iron (ZVI) and sulfur modified iron (SMI) in reducing mobile Tc(VII) to immobile Tc(IV) and iodate incorporation into calcite, the coupled effects from simultaneously using these remedial technologies have not been previously studied. In this first-of-a-kind laboratory study, we used two efficient reductants (i.e., ZVI and SMI) and calcite-forming solutions to simultaneously remove aqueous Tc(VII) and iodate via reduction and incorporation, respectively. The results confirmed that Tc(VII) was rapidly removed from the aqueous phase via reduction to Tc(IV). ZVI removed Tc(VII) faster than SMI, although both had removed the same amount by the end of the experiments. Most of the aqueous iodate was rapidly transformed to iodide, and therefore was not incorporated into calcite, but instead remained in the aqueous phase. The iodate reduction to iodide was much faster than iodate incorporation into calcite, suggesting that this remedial pathway is not efficient in removing aqueous iodate when strong reductants are present. Other experiments suggested that iodate removal via calcite precipitation should occur first and then reductants should be added for Tc(VII) removal. Although ZVI can negatively impact microbial populations and thereby inhibit natural attenuation mechanisms, only changes in the makeup of the microbial community were observed. However, these changes in the microbial community may have an impact on remediation efforts in the long term that could not be seen in a short-term study. The results underscore the importance of identifying interactions between natural attenuation pathways and remediation technologies that only target individual contaminants.

Keywords: Tc removal, groundwater remediation, nuc↗
DOE OSTI2014

Renewability and sustainability aspects of nuclear energy

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2}; 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ∼1.3.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Evaluation of Gas Retention in Waste Simulants: Tall Column Experiments

Gas generation in Hanford’s underground waste storage tanks can lead to gas accumulation within the layer of settled solids (sludge) at the tank bottom. The gas, which typically has hydrogen as the major component together with other flammable species, is formed principally by radiation-driven chemical reactions. Accumulation of these gases within the sludge in a waste tank is undesirable and limits the amount of tank volume for waste storage. Further, accumulation of large amounts of gas in the sludge may potentially result in an unacceptable release of the accumulated gas if the sludge-layer density is reduced to less than that of the overlying sludge or that of the supernatant liquid. Rapid release of large amounts of flammable gases could endanger personnel and equipment near the tank. For this reason, a thorough understanding of the circumstances that can lead to a potentially problematic gas accumulation in sludge layers is needed. To respond to this need, the Deep Sludge Gas Release Event Program (DSGREP) was commissioned to examine gas release behavior in sludges.

Gas Retention↗
DOE OSTI2019

Evaluation of Low-cost, Full HD, CMOS Cameras for Use in the Calcine Retrieval Project - 19530

This paper describes results from radiation testing conducted on low-cost, commercial off-the-shelf, full high-definition (HD) resolution (1920 x 1080 p), complementary metal-oxide-semiconductor (CMOS) camera sensors. The purpose of the testing was to evaluate the sensors' suitability for use in the Calcine Retrieval Project by determining their image quality and life expectancy while operating in an ionizing radiation field of 20.0 Gy/h. The two camera sensors tested were the Axis FA1105 and Axis FA1125. These commercial camera sensors were designed for retail security surveillance and have no radiation-hardened (rad-hard) components. In addition to radiation testing, the sensors' field-of-view, resolution, and low-light image quality were tested at several target distances to ensure they meet project requirements. Testing was divided into two phases. During Phase 1, four FA1105 sensors and four FA1125 sensors were tested simultaneously without any shielding in radiation fields of 25.5 Gy/h and 49.3 Gy/h. In Phase 2, three FA1105 and five FA1125 sensors were shielded by 22.2 mm of lead and tested simultaneously in radiation fields of 26.8 Gy/h, 52.3 Gy/h, 143.0 Gy/h, 200.7 Gy/h, 404.6 Gy/h, and 2,317.5 Gy/h. Dose rate values were established using film dosimetry readings. Apertures in the shielding allowed for a 40-degree field of view while shielding all parts of the sensor except for directly in front of the CMOS sensor array. All testing was conducted at the Idaho Accelerator Center in Pocatello, Idaho, using a 450 keV X-ray generator. Sensor images were automatically captured at regular intervals throughout the testing for qualitative evaluation by the test team. Film dosimetry was placed near the sensors to record actual dose totals. The 450 keV setting generates X-ray radiation with an energy level close to the primary gamma energy level produced by the Cesium-137 isotope in the calcine, which is 662 keV. Results from Phase 1 indicate the sensors failed at an average cumulative dose of 226.3 Gy. For a given 20.0 Gy/h field this results in a mean time to failure (MTTF) of 11.3 hours. Recorded images from all sensors showed minimal image degradation due to radiation-induced noise. When the sensors finally failed, the failure mode was an abrupt loss of video. The image went black suggesting some other component in the sensor head besides the CMOS sensor array was responsible for the failure. During a portion of Phase 1 testing, a commercial, rad-hard camera was irradiated in the 25.5 Gy/h field along with the eight Axis FA series sensors. Image quality from the FA series sensors significantly outperformed the image quality of the rad-hard camera in radiation-induced noise tolerance, image resolution, and dynamic range. During Phase 2 testing, all sensors produced excellent images in radiation fields as high as 404.6 Gy/h. One FA1125 sensor was tested at the highest dose rate of 2,317.5 Gy/h. At this dose rate, the image was color shifted towards red and contained a significant amount radiation-induced noise. The sensor did not fail and was producing usable imagery after one hour, at which time the test was ended due to allotted beam-time at the test facility. Eighteen hours after X-rays were shut off, the sensor image returned to normal color and all eight sensors continued to operate without any visible, permanent image degradation. For Phase 2 testing, it was not possible to establish a MTTF since none of the shielded sensors failed during the allotted test time. However, given the highest cumulative dose of 4,142.1 Gy received by a shielded FA1125 sensor, a minimum life expectancy of 207 hours (8.6 days) is predicted while operating in a 20.0 Gy/h field. Test results indicate the radiation tolerance of the FA series sensor's CMOS array, along with the small form factor, full HD image quality, wide dynamic range, and Ethernet standard makes it suitable for short-term use in monitoring and visually documenting waste remediation efforts for the Calcine Retrieval Project at the INL. The sensor's low-cost of about $140 US, compared to commercial rad-hard cameras costing between $5,000 and $7,000 US each, allow them to be used as an affordable, disposable resource for many short-duration, radioactive waste remediation efforts where decontamination of an expensive, commercial rad-hard camera is undesirable or not possible. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1976

Application of LANDSAT-2 to the management of Delaware's marine and wetland resources

The author has identified the following significant results. The spectral signature of the acid waste disposal plume investigated 38 miles off the Delaware coast, is caused primarily by scattering from particles in the form of suspended ferric iron floc. In comparison, the absorption caused by the dissolved fraction of iron and other substances has a negligible effect on the spectral signature. Ocean waste disposal plumes were observed by LANDSAT-1 and -2 during dump up to 54 hours afer dump during fourteen different passes over the Delaware test site. The spatial resolution, radiometric sensitivity, and spectral band location of the LANDSAT multispectral scanner are sufficient to identify the location of ocean disposal plumes. The movement and dispersion of ocean waste disposal plumes can be estimated if the original dump location, time, and injection method are known. Operating LANDSAT in the high gain mode helps to determine plume dispersion more accurately.

Klemas, V.↗
DOE OSTI2013

TRACKING SURPLUS PLUTONIUM FROM WEAPONS TO DISPOSITION

Supporting nuclear nonproliferation and global security principles, beginning in 1994 the United States has withdrawn more than 50 metric tons (MT) of government-controlled plutonium from potential use in nuclear weapons. The Department of Energy (DOE), including the National Nuclear Security Administration, established protocols for the tracking of this "excess" and "surplus" plutonium, and for reconciling the current storage and utilization of the plutonium to show that its management is consistent with the withdrawal policies. Programs are underway to ensure the safe and secure disposition of the materials that formed a major part of the weapons stockpile during the Cold War, and growing quantities have been disposed as waste, after which they are not included in traditional nuclear material control and accountability (NMC&A) data systems. A combination of resources is used to perform the reconciliations that form the basis for annual reporting to DOE, to U.S. Department of State, and to international partners including the International Atomic Energy Agency.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2012

EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES

Historically, high-level waste (HLW) glasses have been formulated with a low liquideus temperature (T{sub L}), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T{sub 0.01}), nominally below 1050 C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions ({approx} 850 C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES

Historically, high-level waste (HLW) glasses have been formulated with a low liquidus temperature (TL), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T0.01), nominally below 1050°C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions (~ 850°C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

Empirical model↗
DOE OSTI2015

The Use of Underground Research Laboratories to Support Repository Development Programs. A Roadmap for the Underground Research Facilities Network.

Under the auspices of the International Atomic Energy Agency (IAEA), nationally developed underground research laboratories (URLs) and associated research institutions are being offered for use by other nations. These facilities form an Underground Research Facilities (URF) Network for training in and demonstration of waste disposal technologies and the sharing of knowledge and experience related to geologic repository development, research, and engineering. In order to achieve its objectives, the URF Network regularly sponsors workshops and training events related to the knowledge base that is transferable between existing URL programs and to nations with an interest in developing a new URL. This report describes the role of URLs in the context of a general timeline for repository development. This description includes identification of key phases and activities that contribute to repository development as a repository program evolves from an early research and development phase to later phases such as construction, operations, and closure. This information is cast in the form of a matrix with the entries in this matrix forming the basis of the URF Network roadmap that will be used to identify and plan future workshops and training events.

MacKinnon, Robert J.↗
DOE OSTI2013

Key differentiators influencing the choice of robust shielded containers

The NDA's Upstream Optioneering project has examined the potential implications of using an alternative type of waste package and its influence on the waste management lifecycle across the NDA estate. Robust Shielded Containers (RSCs) are radioactive waste packages that provide integral radiological shielding, reducing the need for remote handling. The robustness of the container could remove the need to immobilise waste by cement encapsulation. RSCs are routinely used to package ILW for interim storage in Germany and have fairly recently been considered for use in the UK because RSCs have the potential to simplify the waste management lifecycle and enable risk and hazard reduction. In particular, the current baseline (included in Magnox Limited lifetime plans) follows the strategy set out in the Magnox Operating Decommissioning Programme (MODP), in which Type II MOSAIK and Type VI Ductile Cast Iron Containers (DCICs) are used to package (in non-encapsulated form) most Magnox ILW arising prior to 2050. By considering representative types of Intermediate Level Waste (ILW) and exploring potential packaging options for these wastes, this paper identifies the factors that could differentiate between cases in which RSCs would, or would not, be an appropriate option. The potential role of RSCs across the waste management lifecycle is examined, from retrieval of waste through to emplacement at a Geological Disposal Facility (GDF), including consideration of other potential uses of RSCs, such as temporary storage of raw wastes for which appropriate treatment and conditioning measures have yet to be developed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Multivariate Linear Models for In Situ Monitoring of Low-Activity Nuclear Waste Using Infrared and Raman Spectroscopy - 20218

This paper presents the use of in situ monitoring tools for molecular analysis of non-radioactive species found in the supernatant liquid and dissolvable salt cake of nuclear waste at Hanford. We have selected two in situ spectroscopic techniques that do not require any sample preparation or consumables: infrared (in the form of attenuated total reflectance Fourier transform infrared or ATR-FTIR) and Raman spectroscopy. This work aims to evaluate their performance for liquid waste simulants representative of the Hanford DFLAW (Direct Feed Low-Activity Waste) feed. In addition, we have evaluated the performance of four multivariate linear models that extract concentrations from spectral measurements for a data set that was designed to include key molecular species found in liquid low-activity waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Theory of Carbon Dioxide Absorption by Hanford Waste - 18208

Hanford nuclear waste has a very high pH, and Hanford tanks are ventilated with air. Air has CO{sub 2(g)} in it, which is an acid, so it has some affinity for the waste. The absorption of CO{sub 2} by the waste is important because CO{sub 2} absorption consumes hydroxide at the liquid-air interface, and the hydroxide is important for inhibiting tank corrosion. NaOH is periodically added to some tanks in order to mitigate the surface acidification. The NaOH added then becomes part of the waste and must be treated as waste, increasing the cost of the mission. The Hobbs Equation, developed at the Savannah River site, is used at Hanford to predict CO{sub 2} absorption from the tank ventilation system. In this study, the Hobbs Equation functional form is theoretically evaluated from thermodynamic calculations of CO{sub 2} absorption, and diffusion rates in the waste surface. This study theoretically justifies the empirical equation reported by Hobbs. Thermodynamic calculations performed using the Environmental Simulation Program (ESP) software on a typical double-shell tank waste composition indicate that essentially 100 % of CO{sub 2} that contacts the waste will react with the waste whenever the pH is above 10. In the 1990's, a substantial amount of data was collected on the concentration of CO{sub 2} in the off-gas of ventilated double-shell tanks at Hanford. This historic data was mined in order to determine the quantity of CO{sub 2} absorbed by the waste. The results of this data mining activity determined that tanks below approximately pH 10 (i.e., tank 241-AN-107) will show limited CO{sub 2} absorption, consistent with both the Hobbs model and the theory developed here. Two other tank with higher pH showed contrasting CO{sub 2} absorption. One showed substantial CO{sub 2} absorption, and the other showed minimal CO{sub 2} absorption despite the high waste pH in one tank, presumably because there was minimal contact between the waste surface and most of the air flow. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗